Abstract
The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in order to perform uncertainty analysis on and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is a 15 15 pressurized water reactor (PWR) fuel assembly segment with poison and at full power conditions. A statistical methodology is employed for the uncertainty assessment, where cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB library format are considered as normal random variables. Two libraries were created for such purposes, one based on JENDL-4 data and the other one based on the recently released ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. The uncertainty assessment performed on and macroscopic cross-sections, that is based on JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much larger at the thermal and resonant regions than, for instance, the covariance matrix based on ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different uncertainty assessments.
Highlights
E signi cant increase in capacity of new computational technology made it possible to switch to a newer generation of complex codes, which are capable of representing the feedback between core thermal-hydraulics and neutron kinetics in detail. e coupling of advanced, best estimate (BE) models is recognized as an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines
E nominal solution to this exercise is shown in Tables 2 and 3, where the fast and thermal macroscopic cross-sections and kk∞ are presented, respectively, using libraries based on both JENDL-4 and ENDF/B-VII.1 data
A statistical uncertainty analysis was performed on lattice calculations using the DRAGONv4.05 code. e input uncertainty space corresponded to the microscopic cross-sections of the different nuclides of the DRAGLIB library. is work is one of the rst attempts to process in multigroup format uncertainties from modern nuclear libraries such as JENDL-4 and ENDF/B-VII.1, so they could be applied to the uncertainty assessment of lattice calculations. us, con dence in the results of advance lattice codes can be obtained through the use of a statistical uncertainty analysis
Summary
E signi cant increase in capacity of new computational technology made it possible to switch to a newer generation of complex codes, which are capable of representing the feedback between core thermal-hydraulics and neutron kinetics in detail. e coupling of advanced, best estimate (BE) models is recognized as an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Hébert [2] developed a nuclear data library production system that recovers and formats nuclear data required by the advanced lattice code DRAGON version 4 [3] and higher versions For these purposes, a new postprocessing module known as DRAGR was included in NJOY99, which is capable of creating the so-called DRAGLIB nuclear data library for the DRAGON v 4.05 code. Erefore, the output sample formed by the 450 code calculations infers to cover 95% of the multivariate output population, with at least a 95% of con dence All these is performed twice, once for libraries based on JENDL-4 data and another time for libraries based on ENDF/B-VII. data, respectively, for their further comparison
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